[Federal Register Volume 59, Number 5 (Friday, January 7, 1994)]
[Proposed Rules]
[Pages 979-984]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-341]
[[Page Unknown]]
[Federal Register: January 7, 1994]
VOL. 59, NO. 5
Friday, January 7, 1994
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AC93
Codes and Standards for Nuclear Power Plants; Subsection IWE and
Subsection IWL
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) proposes to amend its
regulations to incorporate by reference the 1992 Edition with the 1992
Addenda of Subsection IWE, ``Requirements for Class MC and Metallic
Liners of Class CC Components of Light-Water Cooled Power Plants,'' and
Subsection IWL, ``Requirements for Class CC Concrete Components of
Light-Water Cooled Power Plants,'' of Section XI, Division 1, of the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code) with specified modifications and a limitation.
Subsection IWE of the ASME Code provides rules for inservice
inspection, repair, and replacement of Class MC pressure retaining
components and their integral attachments and of metallic shell and
penetration liners of Class CC pressure retaining components and their
integral attachments in light-water cooled power plants. Subsection IWL
of the ASME Code provides rules for inservice inspection and repair of
the reinforced concrete and the post-tensioning systems of Class CC
components. Licensees would be required to incorporate Subsection IWE
and Subsection IWL into their routine inservice inspection (ISI)
program. Licensees would also be required to expedite implementation of
the containment examinations and complete the expedited examination in
accordance with Subsection IWE and Subsection IWL within 5 years of the
effective date of this rule. Provisions have been proposed that would
prevent unnecessary duplication of examinations between the expedited
examination and the routine 120-month ISI examinations. Subsection IWE
and Subsection IWL have not been previously incorporated by reference
into the NRC regulations. This proposed amendment would specify
requirements to assure that the critical areas of containments are
routinely inspected to detect defects that could compromise a
containment's pressure-retaining integrity.
DATES: Comment period expires March 23, 1994. Comments received after
this date will be considered if it is practical to do so, but assurance
of consideration cannot be given except as to comments received on or
before this date.
ADDRESSES: Written comments or suggestions may be submitted to the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, Attention: Docketing and Service Branch. Deliver
comments to: 11555 Rockville Pike, Rockville, MD between 7:45 am and
4:15 pm Federal workdays. Copies of the regulatory analysis, the
environmental assessment and finding of no significant impact, the
supporting statement submitted to the Office of Management and Budget,
and comments received may be examined in the Commission's Public
Document Room at 2120 L Street, NW. (Lower Level), Washington, DC.
FOR FURTHER INFORMATION CONTACT: Mr. W.E. Norris, Division of
Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear
Regulatory Commission, Washington, DC 20555, telephone (301) 492-3805,
or Mr. H.L. Graves, Division of Engineering, Office of Nuclear
Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC
20555, telephone (301) 492-3813.
SUPPLEMENTARY INFORMATION:
Background
The NRC is taking the proposed action for the purpose of ensuring
that containments continue to maintain or exceed minimum accepted
design wall thicknesses and prestressing forces as provided for in
industry standards used to design containments (e.g., Section III and
Section VIII of the ASME Code, and the American Concrete Institute
Standard ACI-318), as reflected in license conditions, technical
specifications, and licensee commitments (e.g., the Final Safety
Analysis Report). The NRC also believes enhanced ISI examinations are
needed and are justified to supplement existing requirements specified
in General Design Criterion (GDC) 16, and GDC 53, appendix A to 10 CFR
part 50, and appendix J to 10 CFR part 50. Appendix J requires a
general visual inspection of the containment but does not provide
specific guidance on how to perform the necessary containment
examinations. This has resulted in a large variation with regard to the
performance and the effectiveness of containment inspections. In view
of the increasing rate of occurrences of degradation in containments
and variability of present containment examinations, the NRC has
determined that it is necessary to include more detailed requirements
for the periodic examination of containment structures in the
regulations to assure that the critical areas of containments are
periodically inspected to detect defects that could compromise the
containment's pressure-retaining and leak-tight capability. Recent
changes and additions to the ASME Code include provisions to address
the concerns outlined above. The NRC proposes to make these provisions
mandatory by amending 10 CFR 50.55a to incorporate by reference these
additional portions of the ASME Code (Subsection IWE and Subsection
IWL). Subsection IWE and Subsection IWL have not been previously
incorporated by reference into the NRC's regulations.
The rate of occurrence of corrosion and degradation of containments
has been increasing at operating nuclear power plants. Since 1986,
twenty-one (21) instances of corrosion in steel containments have been
reported. In two cases, thickness measurements of the walls revealed
areas where the wall thickness was at or below the minimum design
thickness. Since the early 1970s, thirty-one (31) incidents of
containment degradation related to post-tensioning systems of concrete
containments have been reported. Four recent additional incidents which
involved grease leakage from tendons have been investigated. In
addition to grease leakage, these incidents showed signs of leaching of
the concrete.
Over one-third of the operating containments have experienced
corrosion or other degradation. Almost one-half of these occurrences
were found by the NRC through its inspections or audits of plant
structures, or by licensees because they were alerted to a degraded
condition at another site. Examples of degradation not found by
licensees, but initially detected at plants through NRC inspections
include: Steel containment shell corrosion in the drywell sand cushion
region (wall thickness reduced to below minimum design thickness);
steel containment shell torus corrosion (wall thickness at or near
minimum design thickness); grease leakage from the tendons of
prestressed concrete containments, and water seepage, as well as
concrete cracking in concrete containments.
There are several GDC criteria and ASME Code sections which
establish minimum requirements for the design, fabrication,
construction, testing, and performance of structures, systems, and
components important to safety in water-cooled nuclear power plants.
Criterion 16, ``Containment design,'' requires the provision of reactor
containment and associated systems to establish an essentially leak-
tight barrier against the uncontrolled release of radioactivity into
the environment and to ensure that the containment design conditions
important to safety are not exceeded for as long as required for
postulated accident conditions. Section III and Section VIII of the
ASME Code, and the American Concrete Institute provide design
specifications for minimum wall thicknesses and prestressing forces of
containments, and these are reflected in license conditions, technical
specifications, and licensee commitments for the operating plants.
Criterion 53, ``Provisions for containment testing and
inspection,'' requires that the reactor containment design permit: (1)
Appropriate periodic inspection of all important areas, such as
penetrations; (2) an appropriate surveillance program; and (3) periodic
testing at containment design pressure of the leak-tightness of
penetrations which have resilient seals and expansion bellows. Appendix
J, ``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors,'' of 10 CFR part 50 contains specific rules for leakage
testing of containments. Paragraph V. A. of appendix J requires that a
general inspection of the accessible interior and exterior surfaces of
the containment structures and components be performed prior to any
Type A test to uncover any evidence of structural deterioration that
may affect either the containment structural integrity or leak-
tightness. (Type A test means tests intended to measure the primary
reactor containment overall integrated leakage rate: (1) After the
containment has been completed and is ready for operation, and (2) at
periodic intervals thereafter). None of these existing requirements,
however, provide specific guidance on how to perform the necessary
containment examinations. This lack of guidance has resulted in a large
variation in licensee containment examination programs, such that there
have been cases of noncompliance with GDC 16. Based on the results of
inspections and audits, as well as plant operational experiences, it is
clear that many licensee containment examination programs have not
detected degradation that could ultimately result in a compromise to
the pressure-retaining capability. Some containment structures have
also been found to have undergone a significant level of degradation
that was not detected by these programs.
The NRC believes that more specific ISI requirements, which expand
upon existing requirements for the examination of containment
structures in accordance with GDC 53 and appendix J , are needed and
are justified for the purpose of ensuring that containments continue to
maintain minimum design wall thicknesses and prestressing forces as
provided for in industry standards used to design containments (e.g.,
Section III and Section VIII of the ASME Code, and the American
Concrete Institute Standard ACI-318), as reflected in license
conditions, technical specifications, and written licensee commitments
(e.g., the Final Safety Analysis Report). There exists a serious
concern, based on actual operating experience, regarding continued
compliance by the operating plants with existing requirements for
ensuring containment minimum design wall thicknesses and prestressing
forces if the proposed action is not taken. The NRC also believes that
the occurrences of corrosion and other degradation discussed above
would have been detected by licensees implementing the comprehensive
periodic examinations set forth in Subsection IWE and Subsection IWL of
the ASME Code proposed for incorporation by reference into 10 CFR
50.55a.
The Nuclear Management and Resources Council (NUMARC) has developed
a number of industry reports to address license renewal issues. Two of
them, one for PWR containments and the other for BWR containments, were
developed for the purpose of managing age-related degradation of
containments on a generic basis. The NUMARC plan for containments
relies on the examinations contained in Subsection IWE and Subsection
IWL to manage age-related degradation, and this plan assumes that these
examinations are ``in current and effective use.'' In the BWR
Containment Industry Report, NUMARC concluded that ``On account of
these available and established methods and techniques to adequately
manage potential degradation due to general corrosion of freestanding
metal containments, no additional measures need to be developed and, as
such, general corrosion is not a license renewal concern if the
containment minimum wall thickness is maintained and verified.''
Similarly, in the PWR Containment Industry Report, NUMARC concluded
that potentially significant degradation of concrete surfaces, the
post-tensioning system, and the liners of concrete containments could
be managed effectively if periodically examined in accordance with the
requirements contained in Subsection IWE and Subsection IWL.
The five modifications, which are contained in one paragraph of the
proposed rule, address two concerns of the NRC. The first concern is
that certain recommendations for tendon examinations that are included
in Regulatory Guide 1.35, Rev. 3, are not addressed in Subsection IWL
(this involves four of the modifications, (ix)(A)-(D)). The ASME Code
has considered these four issues and has adopted them in Subsection
IWL. These issues will be published in future addenda. The second
concern is that if there is visible evidence of degradation of the
concrete (e.g., leaching, surface cracking) there may also be
degradation of inaccessible areas. This fifth modification ((ix)(E))
contains a provision which would require an evaluation of inaccessible
areas when visible conditions exist that could result in degradation of
these areas.
The limitation specifies the 1992 Edition with 1992 Addenda of
Subsection IWE and Subsection IWL as the earliest version of the ASME
Code the NRC finds acceptable. This edition and addenda combination
incorporates the concept of base metal examinations and would provide a
comprehensive set of rules for the examination of post-tensioning
systems. As originally published, Subsection IWE preservice examination
and inservice examination rules focused on the examination of welds.
This weld-based examination philosophy was established in the 1970s as
plants were being constructed. It was based on the premise that the
welds in pressure vessels and piping were the areas of greatest
concern. As containments have aged, degradation of base metal, rather
than welds, has been found to be the issue of concern. The 1991 Addenda
to the 1989 Edition, the 1992 Edition and the 1992 Addenda to Section
XI, Subsection IWE, all have furthered the incorporation of base metal
examinations.
The proposed rulemaking incorporates a provision for an expedited
examination schedule. This expedited examination schedule is necessary
to prevent a delay in the implementation of Subsection IWE and
Subsection IWL (Table 4 of Enclosure 2 lists each plant and the delay
in implementation which would be encountered without an expedited
implementation schedule). Provisions have been incorporated in the
proposed rule so that the expedited examination which would be required
5 years after the effective date of the rule and the routine 120-month
examinations are not duplicated.
The NRC has reviewed the 1992 Edition with the 1992 Addenda of
Subsection IWE and Subsection IWL of Section XI of the ASME Code and
has found that with the specified modifications these subsections of
Section XI address current experience and provide a sound basis for
ensuring the structural integrity of containments. NRC endorsement of
Subsection IWE and Subsection IWL in its regulations would provide a
method of improving containment examination practices by incorporating
rules into the regulatory process that have received industry
participation in their development and acceptance by the NRC.
Existing Sec. 50.55a(g), ``Inservice inspection requirements,''
specifies the requirements for preservice and inservice examinations
for Class 1 (Class 1 refers to components of the reactor coolant
pressure boundary), Class 2 (Class 2 quality standards are applied to
water- and steam-containing pressure vessels, heat exchangers (other
than turbines and condensers), storage tanks, piping, pumps, and valves
that are part of the reactor coolant pressure boundary (e.g., systems
designed for residual heat removal and emergency core cooling)), and
Class 3 (Class 3 quality standards are applied to radioactive-waste-
containing pressure vessels, heat exchangers (other than turbines and
condensers), storage tanks, piping, pumps, and valves (not part of the
reactor coolant pressure boundary)) components and their supports.
Neither Subsection IWE (Class MC--metal containments) nor Subsection
IWL (Class CC--concrete containments) is presently incorporated by
reference into the NRC regulations.
Proposed Sec. 50.55a(g)(4) specifies the containment components to
which the ASME Code Class MC and Class CC inservice inspection
classifications incorporated by reference in this proposed rule would
apply.
Proposed Sec. 50.55a (g)(4) (v)(A), (v)(B), and (v)(C) specify
Subsection IWE and Subsection IWL rules for repairs and replacements of
metal and concrete containments. This is consistent with the long-
standing intent and ongoing application by NRC and licensees to utilize
the rules of Section XI when performing repairs and replacements of
applicable components and their supports.
Proposed Sec. 50.55a(b)(2)(vi) would incorporate a limitation
specifying the 1992 Edition with 1992 Addenda of Subsection IWE and
Subsection IWL as the earliest ASME Code version the NRC finds
acceptable. This edition and addenda combination incorporates the
concept of base metal examinations and provides a comprehensive set of
rules for the examination of post-tensioning systems.
Proposed Sec. 50.55a(b)(2)(ix) would specify five modifications
that must be implemented when using Subsection IWL. Four of these
issues are identified in Regulatory Guide 1.35, Revision 3, but are not
currently addressed in Subsection IWL.
Proposed Sec. 50.55a(g)(4)(v) requires that licensees incorporate
containment examinations as part of their routine 120-month inspection
program. It is recognized that when this rule becomes effective, plants
within 2 years of the end of the 120-month interval may have difficulty
developing and completing the containment examination program in a
timely manner. Therefore, proposed Sec. 50.55a(b)(2)(x) specifies that
licensees with less than 2 years remaining in their present ISI
interval may complete the Subsection IWE and the Subsection IWL
portions of their ISI update within 2 years from the end of the present
ISI interval. This is intended to provide licensees with sufficient
time to develop the initial ISI plan and to facilitate maintenance of
one ISI plan instead of two separate plans (i.e, the current Section XI
ISI plan, and the Subsection IWE and Subsection IWL plan). In order to
further reduce the burden on licensees and NRC staff, the Subsection
IWE and Subsection IWL portions of the ISI plan will not have to be
submitted to the NRC for approval. Licensees may simply retain their
initial Subsection IWE and Subsection IWL plans at the site for audit.
Proposed Sec. 50.55a(g)(6)(ii)(B)(1) would require that licensees
conduct the first containment examinations in accordance with
Subsection IWE and Subsection IWL (1992 Edition with the 1992 Addenda),
modified by proposed Sec. 50.55a(b)(2)(ix) within 5 years of the
effective date of the final rule. This expedited examination schedule
is necessary to prevent possible delays in the implementation of
Subsection IWE by as much as 20 years and Subsection IWL by as much as
15 years. Subsection IWE, Table IWE-2500-1, permits the deferral of
most of the required examinations until the end of the 10-year
inspection interval. Adding the ten years that could pass before some
utilities are required to update their ISI plans, a period of 20 years
could pass before the first examinations would take place. Subsection
IWL is based on a 5-year inspection interval. Adding the possible 10
years before update of existing ISI plans, a period of 15 years could
pass before the examinations were performed by plants that have not
voluntarily adopted the provisions of Regulatory Guide 1.35, Rev. 3.
Expediting implementation of the containment examinations is considered
necessary because of the problems that have been identified at various
plants, the need to establish expeditiously a baseline for each
facility, and the need to identify any existing degradation.
Proposed paragraphs (g)(6)(ii)(B)(2) and (g)(6)(ii)(B)(3) would
each provide a mechanism for licensees to satisfy the requirements of
the routine containment examinations and the expedited examination
without duplication. Paragraph (g)(6)(ii)(B)(2) would permit licensees
to avoid duplicating examinations required by both the periodic routine
and expedited examination programs. This provision is intended to be
useful to those licensees that would be required to implement the
expedited examination during the first periodic interval that routine
containment examinations are required. Paragraph (g)(6)(ii)(B)(3) would
allow licensees to use a recently performed examination of the post-
tensioning system to satisfy the requirements for the expedited
examination of the containment post-tensioning system. This situation
would occur for licensees who perform an examination of the post-
tensioning system using Regulatory Guide 1.35 between the effective
date of this rule and the beginning of the expedited examination.
Submission of Comments in Electronic Format
The comment evaluation process will be improved if each comment is
identified with document title, section heading, and paragraph number
addressed. In addition to the original paper copy, submitters are
encouraged to provide a copy of their letter in an electronic format on
IBM PC compatible 3.5- or 5.25-inch diskettes. Data files should be
provided as WordPerfect documents. ASCII text is also acceptable or, if
formatted text is required, data files should be provided in IBM
Revisable-Form Text/Document Content Architecture (RFT/DCA) format. The
format and version should be identified on the diskette's external
label.
Finding of No Significant Environmental Impact
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
subpart A of 10 CFR part 51, that this rule, if adopted, would not be a
major Federal action significantly affecting the quality of the human
environment and therefore an environmental impact statement is not
required.
This proposed rule is one part of a regulatory framework directed
to ensuring containment integrity. Therefore, in the general sense, the
proposed rule would have a positive impact on the environment. The
proposed rule would incorporate by reference in the NRC regulations
requirements contained in the ASME Code for the inservice inspection of
the containments of nuclear power plants. Actions required of
applicants and licensees to implement the proposed rule are of a
routine nature that should not increase the potential for a negative
environmental impact.
The environmental assessment and finding of no significant impact
on which this determination is based are available for inspection at
the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC. Single copies of the environmental assessment and the
finding of no significant impact are available from Mr. W.E. Norris,
Division of Engineering, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, telephone (301)
492-3805, or Mr. H.L. Graves, Division of Engineering, Office of
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, telephone (301) 492-3813.
Paperwork Reduction Act Statement
This proposed rule amends information collection requirements that
are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et
seq). This rule has been submitted to the Office of Management and
Budget for review and approval of the paperwork requirements.
The public reporting burden for this collection of information is
estimated to average 4,000 hours per response for development of an
initial inservice inspection plan and 10,000 hours per response for the
update of the plan and periodic examinations, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
collection of information. Send comments regarding this burden estimate
or any other aspect of this collection of information, including
suggestions for reducing this burden, to the Information and Records
Management Branch (MNBB-7714), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the Desk Officer, Office of
Information and Regulatory Affairs, NEOB-3019, (3150-0011), Office of
Management and Budget, Washington, DC 20503.
Documented Evaluation
The Commission has prepared a draft summary of documented
evaluation on this proposed regulation. The draft evaluation is
available for inspection in the NRC Public Document Room, 2120 L Street
NW. (Lower Level), Washington, DC. Single copies of the analysis may be
obtained from Mr. W.E. Norris, Division of Engineering, Office of
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, telephone (301)492-3805, or from Mr. H.L. Graves,
Division of Engineering, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, telephone
(301)492-3813.
The Commission requests public comment on the draft summary of
documented evaluation. Comments on the draft evaluation may be
submitted to the NRC as indicated under the ADDRESSES heading.
Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission hereby certifies that this rule will not, if
promulgated, have a significant economic impact on a substantial number
of small entities. This proposed rule affects only the operation of
nuclear power plants. The companies that own these plants do not fall
within the scope of the definition of ``small entities'' set forth in
the Regulatory Flexibility Act or the Small Business Size Standards set
out in regulations issued by the Small Business Administration at 13
CFR part 121. Since these companies are dominant in their service
areas, this proposed rule does not fall within the purview of the Act.
Backfit Statement
The NRC is taking the proposed action for the purpose of ensuring
that containment structures continue to maintain or exceed minimum
accepted design wall thicknesses and prestressing forces as provided
for in industry standards used to design containment structures, as
reflected in license conditions, technical specifications, and licensee
commitments. Therefore, under 10 CFR 50.109(a)(4)(i) a backfit analysis
need not be prepared for this rule. A summary of the documented
evaluation required by Sec. 50.109(a)(4) to support this conclusion is
set forth below.
GDC 16, ``Containment design,'' requires the provision of reactor
containment and associated systems to establish an essentially leak-
tight barrier against the uncontrolled release of radioactivity into
the environment and to ensure that the containment design conditions
important to safety are not exceeded for as long as required for
postulated accident conditions.
Criterion 53, ``Provisions for containment testing and
inspection,'' requires that the reactor containment design permit: (1)
Appropriate periodic inspection of all important areas, such as
penetrations; (2) an appropriate surveillance program; and (3) periodic
testing at containment design pressure of the leak-tightness of
penetrations which have resilient seals and expansion bellows. Appendix
J, ``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors,'' of 10 CFR part 50 contains specific rules for leakage
testing of containments. Paragraph V. A. of appendix J requires that a
general inspection of the accessible interior and exterior surfaces of
the containment structures and components be performed prior to any
Type A test to uncover any evidence of structural deterioration that
may affect either the containment structural integrity or leak-
tightness (Type A test means tests intended to measure the primary
reactor containment overall integrated leakage rate: (1) After the
containment has been completed and is ready for operation, and (2) at
periodic intervals thereafter). None of these existing requirements,
however, provide specific guidance on how to perform the necessary
containment examinations. This lack of guidance has resulted in a large
variation in licensee containment examination programs, such that there
have been cases of noncompliance with GDC 16. Based on the results of
inspections and audits, and plant operational experiences, it is clear
that many licensee containment examination programs have not detected
degradation that could result in a compromise of pressure-retaining
capability. The location and extent of corrosion or degradation in a
containment can be critical to the containment's behavior during an
accident.
The metal containment structure of operating nuclear power plants
were designed in accordance with either Section III, Subsection NE,
``Class MC Components,'' or Section VIII, of the ASME Code. These
subsections contain provisions for the design and construction of metal
containment structures, including methods for determining the minimum
required wall thicknesses. The minimum wall thickness is determined so
that the metal containment structure will continue to maintain its
structural integrity under the various stressors and degradation
mechanisms which act on it.
The American Concrete Institute Standard ACI-318 contains
provisions for designing and constructing the post-tensioning systems
of concrete containment structures, including methods for determining
the prestressing forces. The post-tensioning system is designed so that
the concrete containment structure will continue to maintain its
structural integrity under the various stressors and degradation
mechanisms which act on it.
These requirements for minimum design wall thicknesses and
prestressing forces as provided in these industry standards used to
design containment structures are reflected in license conditions,
technical specifications, and licensee commitments (e.g., the Final
Safety Analysis Report).
The rate of occurrence of corrosion and degradation of containment
structures has been increasing at operating nuclear power plants. Over
one-third of operating containment structures have experienced
corrosion or other degradation. Almost one-half of the occurrences were
first identified by the NRC through its inspections or structural
audits, or by licensees because they were alerted to a degraded
condition at another site. Examples of degradation not found by
licensees, but initially detected at plants through NRC inspections
include (1) corrosion of steel containment shells in the drywell sand
cushion region, resulting in wall thickness reduced to below the
minimum design thickness; (2) corrosion of the torus of the steel
containment shell (wall thickness at or near minimum design thickness);
(3) grease leakage from the tendons of prestressed concrete
containments; and (4) water seepage, as well as concrete cracking in
concrete containments.
The NRC believes that more specific ISI requirements, that expand
upon existing requirements for the examination of containment
structures in accordance with GDC 53, and appendix J are needed and are
justified to ensure that containment structures continue to maintain or
exceed minimum accepted design wall thicknesses and prestressing forces
as reflected in license conditions, technical specifications, or
licensee commitments. Based on actual operating experience, a serious
concern exists regarding continued compliance by the operating plants
with existing requirements for ensuring containment minimum design wall
thicknesses and prestressing forces if the proposed action is not
taken. The NRC also believes that the occurrences of corrosion and
other degradation discussed above would have been detected by licensees
when conducting the comprehensive periodic examinations set forth in
Subsection IWE and Subsection IWL of the ASME Code, as proposed for
incorporation by reference into 10 CFR 50.55a.
Recent changes and additions to the ASME Code include provisions to
address the concerns outlined above; and the staff proposes to make
these provisions mandatory by amending 10 CFR 50.55a to incorporate by
reference these additional portions of the ASME Code (Subsection IWE
and Subsection IWL). The Commission concludes that this proposed
backfit is necessary to ensure compliance with GDCs 16 and 53, appendix
J, minimum design wall thicknesses in metal containments, and the
prestressing forces of concrete containments, which are applicable to
all licensees through license conditions, technical specifications, and
licensee commitments.
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal Penalties, Fire
protection, Incorporation by reference, Intergovernmental relations,
Nuclear power plants and reactors, Radiation protection, Reactor siting
criteria, Reporting and recordkeeping requirements.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended, the Energy Reorganization
Act of 1974, as amended, and 5 U.S.C. 533, the NRC is proposing to
adopt the following amendments to 10 CFR part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for part 50 continues to read as follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,
83 Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101,
185, 68 Stat. 936, 955, as amended (42 U.S.C. 2131, 2235); sec. 102,
Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13,
50.54(dd) and 50.103 also issued under sec. 108, 68 Stat. 939, as
amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56
also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections
50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L.
91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also
issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections
50.58, 50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat.
2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68
Stat. 939 (42 U.S.C. 2152). Sections 50.80-50.81 also issued under
sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also
issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
2. Section 50.55a is amended by adding paragraphs (b)(2)(vi),
(b)(2)(ix), (b)(2)(x), (g)(4)(v), and (g)(6)(ii)(B), and revising the
introductory text of paragraph (g)(4) to read as follows:
Sec. 50.55a Codes and standards.
* * * * *
(b) * * *
(2) * * *
(vi) Effective edition and addenda of Subsection IWE and Subsection
IWL, Section XI. When using Subsection IWE and Subsection IWL, the 1992
Edition with the 1992 Addenda is the only acceptable Edition and
Addenda.
* * * * *
(ix) Examination of concrete containments.
(A) All grease caps that are accessible must be visually examined
to detect grease leakage or grease cap deformations. Grease caps must
be removed for this examination when there is evidence of grease cap
deformation that indicates deterioration of anchorage hardware.
(B) An Engineering Evaluation Report must be prepared as prescribed
in IWL-3300(a), (b), (c), and (d) when evaluation of consecutive
surveillances of prestressing forces for the same tendon or tendons in
a group indicates a trend of prestress loss such that the tendon
force(s) would be less than the minimum design prestress requirements
before the next inspection interval.
(C) When the elongation corresponding to a specific load (adjusted
for effective wires or strands) during retensioning of tendons differs
by more than 10 percent from that recorded during the last measurement,
an evaluation must be performed to determine whether the difference is
related to wire failures or slip of wires in anchorages. A difference
of more than 10 percent must be identified in the ISI Summary Report.
(D) The licensee shall identify the following conditions, if they
occur, in the ISI Summary Report:
(1) The sampled sheathing filler grease contains chemically
combined water exceeding 10 percent by weight or the presence of free
water;
(2) The absolute difference between the amount removed and the
amount replaced may not exceed 10 percent of the tendon net duct
volume.
(3) Grease leakage is detected during general visual examination of
the containment surface.
(E) The licensee shall evaluate the acceptability of inaccessible
areas when conditions exist in accessible areas that could indicate the
presence of or result in degradation to such inaccessible areas. For
each inaccessible area identified, the licensee shall provide the
following in the ISI Summary Report:
(1) A description of the type and estimated extent of degradation,
and the conditions that led to the degradation;
(2) An evaluation of each area, and the result of the evaluation,
and;
(3) A description of necessary corrective actions.
(x) Subsection IWE and Subsection IWL inservice inspection plans.
Licensees that have less than 2 years remaining in their present 120-
month inservice inspection interval on (effective date of the final
rule) may defer completion of the Subsection IWE and Subsection IWL
portions of the inspection plan for the next 120-month inspection
interval for up to 2 years from the end of the present interval.
* * * * *
(g)* * *
(4) Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, components (including supports) which
are classified as ASME Code Class 1, Class 2, and Class 3 must meet the
requirements, except design and access provisions and preservice
examination requirements, set forth in Section XI of editions of the
ASME Boiler and Pressure Vessel Code and Addenda that become effective
subsequent to editions specified in paragraphs (g)(2) and (g)(3) of
this section and are incorporated by reference in paragraph (b) of this
section, to the extent practical within the limitations of design,
geometry and materials of construction of the components. Components
which are classified as Class MC pressure retaining components and
their integral attachments, and components which are classified as
Class CC pressure retaining components and their integral attachments
must meet the requirements, except design and access provisions and
preservice examination requirements, set forth in Section XI of the
ASME Boiler and Pressure Vessel Code and Addenda that are incorporated
by reference in paragraph (b), subject to the limitation listed in
paragraph (b)(2)(vi) and the modifications listed in paragraphs
(b)(2)(ix) and (b)(2)(x) of this section, to the extent practical
within the limitations of design, geometry and materials of
construction of the components.
* * * * *
(v) For a boiling or pressurized water-cooled nuclear power
facility whose construction permit was issued after January 1, 1956:
(A) Metal containment pressure retaining components and their
integral attachments must meet the inservice inspection, repair, and
replacement requirements applicable to components which are classified
as ASME Code Class MC;
(B) Metallic shell and penetration liners which are pressure
retaining components and their integral attachments in concrete
containments must meet the inservice inspection, repair, and
replacement requirements applicable to components which are classified
as ASME Code Class CC; and
(C) Concrete containment pressure retaining components and their
integral attachments, and the post-tensioning systems of concrete
containments must meet the inservice inspection and repair requirements
applicable to components which are classified as ASME Code Class CC.
* * * * *
(6)* * *
(ii)* * *
(B) Expedited examination of containment.
(1) Licensees of all operating nuclear power plants shall implement
the examinations specified for the first inspection interval in
Subsection IWE and Subsection IWL of the 1992 Edition with the 1992
Addenda in conjunction with the modifications specified in Sec. 50.55a
(b)(2)(ix) by (a date will be inserted that is 5 years later than the
effective date of the final rule).
(2) The expedited examination may be used to satisfy the
requirements of routinely scheduled examinations of Subsection IWE
subject to IWA-2430(c) when the expedited examination occurs during the
first containment inspection interval.
(3) The requirement for the expedited examination of the
containment post-tensioning system may be satisfied by written
commitments that are in place before (the effective date of the final
rule) for examinations of the post-tensioning system.
* * * * *
Dated at Rockville, Maryland, this 3d day of January 1994.
For the Nuclear Regulatory Commission.
Samuel J. Chilk,
Secretary of the Commission.
[FR Doc. 94-341 Filed 1-6-94; 8:45 am]
BILLING CODE 7590-01-P
